Zirconium alloys have been used as a fuel cladding material for a long time because they have good corrosion resistance and irradiation stability in a reactor environment. However, it was recognized that the severe oxidation and creep ballooning behaviors of zirconium alloy cladding tubes at an accident condition of a nuclear reactor have to be considered to improve nuclear plant safety after t...

Fuel performance models have been developed to assess the performance of internally and externally cooled LWR annular fuel. Such fuel may be operated at 30-50% higher core power density than the current operating LWRs, and to a burnup of 80-100MWd/kgU. The models are used to optimize the fuel design so that it is able to achieve high power density and high burnup, and to identify the features o...

Stability has been the main concern in the development stage of BWRs in history. Although stability studies of BWRs have a long history and there is extensive literature available, the literature on fuel performance under unstable power oscillations is very limited. It is very desirable to investigate the fuel integrity in BWR power oscillation conditions. This goal of this work is to examine t...

At the completion of the fuel drying process, used fuel Zry4 cladding typically exhibits a significant population of d-hydride inclusions. These inclusions are in the form of small platelets that are generally oriented both circumferentially and radially within the cladding material. There is concern that radiallyoriented hydride inclusions may weaken the cladding material and lead to issues du...

Presently, LWR (light-water reactor) fuel cladding (tube) is made from Zr-base alloys (e.g., Zircaloy). However, in steam at temperatures above 1000 K (727 °C), these alloys oxidize rapidly, stripping oxygen from water molecules and producing hydrogen, which may cause explosion. Moreover, Zr-base alloys are prone to GTRF (“grid-to-rod” fretting) failure, in which flowinduced vibration of fuel r...

Fuel rod analysis program transient (FRAPTRAN) code was used to study the fuel rod performance during a postulated large break loss of coolant accident (LBLOCA) in Maanshan nuclear power plant (NPP). Previous transient results from thermal hydraulic code, TRACE, with the same LBLOCA scenario, were used as input boundary conditions for FRAPTRAN. The simulation results showed that the peak claddi...